Revealing the enhanced corrosion resistance of Cr/FeCrAl-coated Zr alloy in simulated PWR water
The long-term corrosion behavior and oxide evolution of Cr/FeCrAl-coated Zr-4 alloy in simulated PWR primary water (360 °C and 18.6 MPa) were systematically investigated. Results showed that a continuous triplex oxide film, featuring FeCr 2 O 4 spinel, Fe 3 O 4 spinel and (Cr,Al) 2 O 3 , was developed on the coating, effectively protecting Zr substrate from corrosion. In the early stage, sustained outward Fe/Cr diffusion precipitated on the surface to form the outer spinel layer, while the internal (Cr,Al) 2 O 3 layer grew inwards through oxidant transport. Over time, the diffusion barrier effect of dense (Cr,Al) 2 O 3 layer shifted spinel growth mechanism to dissolution-reprecipitation, accompanied by a decelerated oxide growth. The superior corrosion resistance was dominated by the (Cr,Al) 2 O 3 layer. During exposure, the FeCrAl-Cr interface and the developed near-equiaxed Cr nanograins arrested the vertical cracks along GBs and promoted blunting/deflection, mitigating coating spallation and preserving structural integrity. • The FeCrAl–Cr interface and near-equiaxed Cr nanograins arrest GB-aligned cracks and promote their blunting/deflection. • The Cr/FeCrAl coating enhances corrosion resistance of the Zr substrate in simulated PWR water. • Corrosion mechanism for the Cr/FeCrAl-coated Zr alloy in high-temperature water is proposed. • Continuous (Cr,Al)2O3 layer plays the dominant role in ensuring long-term protection against corrosion.
Authors
- Weijiu Huang
- Jiyuan Cui
- Xiaoling Yang (ORCID: https://orcid.org/0000-0002-1673-7299)
- Baifeng Luan
- Haibo Ruan
- Kazuhiro Ogawa
- Lijun Chen
- Kun Zhang
Institutions
- Chongqing University (CN)
- Tohoku University (JP)
- China General Nuclear Power Corporation (China) (CN)
- Henan Energy & Chemical Industry Group (China) (CN)
- Chongqing University of Arts and Sciences (CN)
Publication Details
- Journal
- Corrosion Science
- Published
- 2026-01-20
- DOI
- https://doi.org/10.1016/j.corsci.2026.113644
- Primary Topic
- Nuclear Materials and Properties
- Type
- article
- Field-Weighted Citation Impact
- 0.00